1. Field of the Invention
This invention relates in general to the field of power producing nuclear reactors and in particular to methods and apparatus for improving neutron flux reduction factors outboard of the core periphery.
2. Description of the Prior Art
It is well known that nuclear reactors are both a technical and commercial success. In one type of commercial nuclear power reactors, commonly referred to as a pressurized light water reactor, a reactive region commonly referred to as a nuclear core contains a nuclear fuel such as uranium 235, as well as other fissile materials, which undergo sustained fission reactions and in so doing, generate heat. There are, of course, other materials in the nuclear core, the presence of such other materials, however, is not germane to this invention and, accordingly, will not be discussed. Typically, a group of mechanical components, which are known as reactor internals, structurally support the core within a heremetically sealed pressure vessel. The reactor internals also direct the flow of a cooling medium, such as light water in pressurized, light water nuclear reactors, into the pressure vessel through the nuclear core, and out of the pressure vessel. The cooling medium, which is alternatively called the reactor coolant, removes the heat generated by the fissioning of the nuclear fuel and transfers the heat to another cooling medium within heat exchangers which are typically located external of the pressure vessel. The second cooling medium, which is usually water, is converted into steam in the heat exchangers and is thereafter used to produce electricity by conventional steam turbine-electrical generator combinations.
The nuclear core, in the type of nuclear reactor described herein, usually comprises, an array of fuel assemblies stacked together in a side-by-side parallel arrangement to form a shape approximately that of a right solid circular cylinder. Each of such fuel assemblies include a multiplicity of elongated fuel rods and control rod guide tubes held together in a parallel array by grids spaced along the fuel assembly length. Each fuel rod may comprise an elongated slender hollow tube to be filled with nuclear fuel pellets and sealed at each end. Top and bottom nozzles on opposite ends of the fuel assembly and secured to the guide tubes provide for reactor coolant flow into and out of the fuel assemblies. The guide tubes allow for the insertion of elongated control rod assemblies into the nuclear core and dispersed among the nuclear fuel. The control rod assemblies provide for reactor control and serve to accomplish other neutronic purposes.
The reactor internals may include a core barrel comprising an elongated cylinder which is interposed between the nuclear core and the cylindrical wall of the pressure vessel. The nuclear core then is positioned within the core barrel. Typically, the reactor coolant enters the pressure vessel through one or more inlet nozzles, flows downward between the pressure vessel and the outside of the core barrel, turns 180.degree., and flows upward through the core and through the space between the outside of the core and the inside of the core barrel. The heated reactor coolant then turns 90.degree. and exits the pressure vessel through one or more exit nozzles and then to the heat exchangers previously mentioned.
In the pressurized light water reactors, such as the one described, the fissioning of the nuclear fuel results from the capture of a neutron by the nucleus of the atoms of the nuclear fuel. It is well known that each neutron producing a fission causes heat and the production of more than one other neutron (on the average 2.1 neutrons are released per capture). To sustain the nuclear chain reaction, at least one of the newly produced neutrons must then fission another atom of fuel. Since the neutrons generated are fast neutrons, and fissioning is enhanced by slow neutrons, it is advantageous that the fast neutrons be slowed down or thermalized within the confines of the nuclear core. The light water reactor coolant is an excellent moderator of neutrons; hence, in reactors primarily using U-235 as the nuclear fuel, it is the primary means by which the fast neutrons produced by the fission process are thermalized or slowed down so as to increase the probability that another fission may occur and thereby sustain the chain reaction. The excess neutrons produced by the fissioning of an atom and not used to fission another atom are accounted for in a number of different ways. Some are absorbed by the reactor internals. Others are slowed down and absorbed by a nuclear poison such as boron which is dissolved in the primary coolant. Other neutrons are absorbed by load follow control rods containing nonburnable control poisons which control rods comprise the means for controlling the nuclear reactor. Others are absorbed by special control rods which are interspersed throughout the nuclear core and made of materials specifically selected to absorb neutrons such as burnable poisons which as their name implies are burned during reactor operation and, therefore, become less effective in proportion to the continually reducing reactivity of the nuclear core. Still other neutrons are absorbed by poisons which build up within the nuclear fuel and are caused by the fission process itself.
Quite obviously, the accounting for the excess neutrons is a complicated matter which can, however, be summarized by stating that some excess neutrons are purposefully absorbed while the remainder are inadvertently absorbed. And, it is desirable to reduce the number that are inadvertently absorbed.
In order to extend the life of the nuclear core as long as is practical so as to minimize time consuming reactor shutdowns for refueling purposes, the fuel assemblies may be provided with enriched nuclear fuel, usually enriched uranium 235. This excessive amount of reactivity is designed into the core at startup so that as the reactivity is depleted over the life of the core, the excess reactivity is then used, thereby extending the life of the core. The amount of enrichment continuously decreases as the reactor operates until such time as the core can no longer sustain the chain reaction. Then the reactor must be shut down and refueled. During the initial stages of reactor operation or during the phase which is known as beginning of life, special neutron absorbing control rods may be inserted within the core and/or additional soluble poisons may be dissolved within the reactor coolant and/or burnable poisons may be included within the fuel assemblies to absorb the excess reactivity. As the excess reactivity decreases due to the nuclear fuel being burned, the amount of insertion of the special control rods and/or the amount of soluble poison and/or the burnable poisons within the special control rods and/or the fuel assemblies are consumed consistent with the reduction in excess reactivity to maintain the chain reaction. In this manner, the excess reactivity is held in abeyance until it is needed.
Enriched uranium is extremely expensive. It is preferable, therefore, to reduce the amount of enrichment whenever possible but without reducing the extended operating length of the life of the core. One recognized method to accomplish this result is by making more efficient use of the neutrons produced by the fission process. In general, such techniques are classified as fuel management techniques. Throughout the years of successful reactor operation, a significant amount of experience has been gained in the area of fuel management. As a result, a three-phase core loading plan has been developed whereby the core is divided into three regions, with each region receiving either new, once burned or twice burned fuel. After a period of reactor operation, the most burned fuel is removed and replaced by the now twice burned fuel which in turn is replaced by the now once burned fuel which is replaced by new fuel. Such a fuel management technique has been termed as out-in-in fuel management, which as its name implies moves fuel radially from the outer core regions progressively inward and toward the center of the core.
The out-in-in fuel management technique has been gradually replaced with a more economical low leakage loading pattern. While more complicated than the out-in-in pattern, the low leakage pattern significantly improves the neutron economy of fuel reload cycles and has thereby reduced the fuel cycle cost for a given energy output. The low leakage loading fuel management technique is designed to minimize the leakage of fast neutrons from out of the core so that these neutrons, as previously explained, may be used for fissioning purposes. Time and reduced fuel cycle costs have verified the success of the low leakage loading pattern.
As an adjunct to the lowering of fuel costs from the use of the low leakage loading pattern, it has been determined that this pattern also results in improved flux reduction factors. That is, the high energy neutron flux which radially emanates from the nuclear core and which may ultimately irradiate the pressure vessel walls is reduced by the low leakage loading pattern. Needless to say, such an effect is obviously advantageous whether achieved by design or otherwise. One complication with the low leakage loading pattern is that it may or may not address a limiting region of the pressure vessel. Indeed, it is even possible that a particular low leakage loading pattern which is designed only to effectuate fuel cycle cost savings may have no effect at all or may even exacerbate neutron irradiation of a particular limiting pressure vessel location such as a weld.
Another complication with using the low leakage loading pattern to improve the flux reduction factor is that the improvement may not be sufficient. Indeed, calculations have been made which show this to be the case. In other words, the low leakage loading pattern may not only be a hit or miss situation but does not sufficiently reduce the irradiation of the pressure vessel walls by high energy flux. Where the loading pattern was adjusted in order to specifically achieve adequate flux reductions, it created unacceptable component thermal margins and adversely affected loss of coolant analysis margins.
Another ostensible solution to reduce pressure vessel neutron fluence is to replace certain of the peripheral fuel assemblies around the core with rods made, for example, of stainless steel. This plan would remove those fuel assemblies contributing to the irradiation of the pressure vessel and add a neutron absorber material between the core and the pressure vessel. Such a solution is not acceptable because of the corresponding reduction in core power rating and, therefore, lower the power output by the power plant.
Another inappropriate solution would be to place additional shielding between the core periphery and the pressure vessel. This solution is unsatisfactory because of space restrictions, requires extensive mechanical redesign and neutronic evaluation of the region outboard of the core, is costly, and requires too much time to effectuate.
Reducing the power output by the peripheral fuel assemblies to sufficiently reduce the fast flux level will cause an unacceptable power reduction in the peripheral fuel region in order to achieve a localized fast flux reduction. To maintain the same total core power output, the inboard region power must increase substantially. This creates unacceptable component thermal margins and adversely affects loss of coolant analysis margins. Hence, this is not a viable solution.
Accordingly, a primary object of the present invention is to provide a method and apparatus to achieve relatively localized fast flux reductions at the core periphery in both the axial and circumferential directions.
Another primary object of the present invention is to provide such localized fast flux reductions on a retrofittable basis; that is, to provide that the method and apparatus may be used with presently built and/or operating nuclear power reactors.
Another primary object of the present invention is to achieve the fast flux reduction without materially adversely affecting core rating and/or power output and/or reactor shutdown margins and/or core thermal margins.
Still another primary object of the present invention is to achieve the fast flux reduction without materially adversely affecting the fuel cycle costs.
The above objects as well as others which are apparent from a reasonable reading and interpretation of this specification, and although they may not be specifically mentioned are all intended to be included within the scope of the invention provided herein.